Description:
Irradiated spent fuel rods, rod sections, and other loose fuel were retrieved from various storage locations at the Battelle hot cells, packaged in stainless steel tubes, and inserted in a new basket assembly in preparation for shipment to EG&G Idaho. Few assemblies Connecticut Yankee S004 and Turkey Point 817 were also retrieved and prepared for shipment. All three fuel assemblies were loaded in shipping cask TN8-L and shipped to EG&G Idaho for storage.

Description:
The primary purpose of this report is to describe the strategy for coupling three process level models to produce an integrated Used Fuel Degradation Model (FDM). The FDM, which is based on fundamental chemical and physical principals, provides direct calculation of radionuclide source terms for use in repository performance assessments. The G-value for H2O2 production (Gcond) to be used in the Mixed Potential Model (MPM) (H2O2 is the only radiolytic product presently included but others will be added as appropriate) needs to account for intermediate spur reactions. The effects of these intermediate reactions on [H2O2] are accounted for in the Radiolysis Model (RM). This report details methods for applying RM calculations that encompass the effects of these fast interactions on [H2O2] as the solution composition evolves during successive MPM iterations and then represent the steady-state [H2O2] in terms of an “effective instantaneous or conditional” generation value (Gcond). It is anticipated that the value of Gcond will change slowly as the reaction progresses through several iterations of the MPM as changes in the nature of fuel surface occur. The Gcond values will be calculated with the RM either after several iterations or when concentrations of key reactants reach threshold values determined from previous sensitivity runs. Sensitivity runs with RM indicate significant changes in G-value can occur over narrow composition ranges. The objective of the mixed potential model (MPM) is to calculate the used fuel degradation rates for a wide range of disposal environments to provide the source term radionuclide release rates for generic repository concepts. The fuel degradation rate is calculated for chemical and oxidative dissolution mechanisms using mixed potential theory to account for all relevant redox reactions at the fuel surface, including those involving oxidants produced by solution radiolysis and provided by the radiolysis model (RM). The RM calculates the ...

Description:
The major component of the DOE spent nuclear fuel inventory is the metallic fuel stored at the Hanford site in the southeastern part of the state of Washington. Most of this fuel was discharged from the N-Reactor; a small part of the inventory is fuel from the early Hanford production reactors. The U.S. Department of Energy (DOE) plans to remove these fuels from the spent fuel storage pools in which they are presently stored, dry them, and place them in interim storage at a location at the Hanford site that is far removed from the Columbia River. It is not yet certain that these fuels will be acceptable for disposal in a mined geologic repository without further treatment, due to their potential pyrophoric character. A practical method for treatment of the Hanford metallic spent fuel, based on an electrorefining process, has been developed and has been demonstrated with unirradiated N-Reactor fuel and with simulated single-pass reactor (SPR) spent fuel. The process can be operated with any desired throughput rates; being a batch process, it is simply a matter of setting the size of the electrorefiner modules and the number of such modules. A single module, prototypic of a production-scale module, has been fabricated and testing is in progress at a throughput rate of 150 kg (heavy metal) per day. The envisioned production version would incorporate additional anode baskets and cathode tubes and provide a throughput rate of 333 kgHM/day. A system with four of these modules would permit treatment of Hanford metallic fuels at a rate of at least 250 metric tons per year.

Description:
The CPP-603 Underwater Fuel Storage Facility (UFSF) Site Integrated Stabilization Management Plan (SISMP) has been developed to describe the activities required for the relocation of spent nuclear fuel (SNF) from the CPP-603 facility. These activities are the only Idaho National Engineering Laboratory (INEL) actions identified in the Implementation Plan developed to meet the requirements of the Defense Nuclear Facilities Safety Board (DNFSB) remediation in the Defense Nuclear Facilities Complex. To date, 622 spent nuclear fuel units have been moved from the CPP-603 north and middle water basins, leaving 743 units in the south basin to be relocated from the facility by December 31, 2000. Besides moving fuels from the CPP-603, in 1993 and 1994 more than 300 fuel storage yokes in the north and middle basins were redundantly rigged because of corrosion problems. More than 200 fuel transfers within the north and middle basins were also made to ensure proper spacing of the fuels, and 104 corroded cans containing spent space reactor fuel were repackaged underwater to prevent potential release of their contents. This document is provided to address the relocation activities for the remaining 743 units in the south basin into wet storage pools at building CPP-666 or into dry storage at the Irradiation Fuel Storage Facility (IFSF).

Description:
In this information handbook, the staff of the U.S. Nuclear Regulatory Commission describes (1) background information regarding the licensing and history of independent spent fuel storage installations (ISFSIs), (2) a discussion of the licensing process, (3) a description of all currently approved or certified models of dry cask storage systems (DCSSs), and (4) a description of sites currently storing spent fuel in an ISFSI. Storage of spent fuel at ISFSIs must be in accordance with the provisions of 10 CFR Part 72. The staff has provided this handbook for information purposes only. The accuracy of any information herein is not guaranteed. For verification or for more details, the reader should refer to the respective docket files for each DCSS and ISFSI site. The information in this handbook is current as of September 1, 1996.

Description:
This document describes the Quality Assurance Program Plan (QAPP) for the Spent Nuclear Fuels (SNF) Canister Storage Building (CSB) Project. The purpose of this QAPP is to control project activities ensuring achievement of the project mission in a safe, consistent and reliable manner.

Description:
Data Quality Objectives (DQOS) for gas and liquid sampling from the sealed canisters in K West Basin have been developed and are presented in this document. The scope of this document is limited primarily to the initial sampling effort. This sampling campaign either supports the selection of canisters to provide fuel for hotcell examinations, supports the demonstration of sampling equipment capabilities or provides an initial assessment of gas/liquid chemistry for comparison to the results of fuel element hotcell examinations. No sampling of canisters has occurred since 1983. It is proposed here that samples of gas and water be analyzed for constituents such as cesium, fission gases, and hydrogen which are markers for corrosion of uranium in a water environment. These data will allow an assessment of the risks involved when particular canisters are opened to retrieve fuel. This sampling campaign will also ensure that canisters with some failed fuel elements are included in the population that is opened for retrieval of fuel for hotcell examinations. Additionally, valuable correlations between the macroscopic visible condition of fuel, hotcell examinations, and the gases generated in canisters will be possible. The analysis of other chemical species in the gas and liquid will allow assessments of the performance of the previously added corrosion inhibitor and possibly assessments of radiolysis. Sampling of canisters will be performed with equipment that opens the valves in the canister lid and draws a 15 ml sample of either gas or water. This work will most likely be performed in one of the pits-associated with the K West Basin.

Description:
The purpose of this document is to provide the design support information for the Vertical Fuel Handling Tools, developed for the removal of N Reactor fuel elements from their storage canisters in the K Basins storage pool and insertion into the Single Fuel Element Can for subsequent shipment to a Hot Cell for examination. Examination of these N Reactor fuel elements is part of the overall characterization effort. These new hand tools are required since previous fuel movement has involved grasping the fuel in a horizontal position. These tools are required to lift an element vertically from the storage canister. Additionally, a Mark II storage canister Lip Seal Protector was designed and fabricated for use during fuel retrieval. This device was required to prevent damage to the canister lip should a fuel element accidentally be dropped during its retrieval, using the handling tools. Supporting documentation for this device is included in this document.

Description:
Onsite and site boundary radiological consequences were estimated for a postulated accidental drop of an Interim Storage Cask (ISC) loaded 7 assemblies at the maximum available burnup. The postulated cask drop was assumed to occur from the maximum physically attainable height during crane movement of the cask. The resulting onsite and site boundary doses of 45 mSv and 0.04 mSv are far below the corresponding 1 Sv and 250 mSv risk guidelines for highly unlikely accidents

Description:
ORIGEN2 calculations were performed for TRIGA spent fuel elements from the Hanford Neutron Radiography Facility. The calculations support storage and disposal and results include mass, activity,and decay heat. Comparisons with underwater dose-rate measurements were used to confirm and adjust the calculations.

Description:
The techniques used for actinide-only burnup credit isotopic validation and criticality validation are presented and discussed. Trending analyses have been incorporated into both methodologies, requiring biases and uncertainties to be treated as a function of the trending parameters. The isotopic validation is demonstrated using the SAS2H module of SCALE 4.2, with the 27BURNUPLIB cross section library; correction factors are presented for each of the actinides in the burnup credit methodology. For the criticality validation, the demonstration is performed with the CSAS module of SCALE 4.2 and the 27BURNUPLIB, resulting in a validated upper safety limit.

Description:
Twelve tests were performed at LLNL to assess loading conditions on a spent fuel casts for side drops, end drops and tipover events. The tests were performed with a 1/3-scale model concrete pad to benchmark the structural analysis code DYNA3D. The side drop and tipover test results are discussed in this report. The billet and test pad were modified with DYNA3D using material properties and techniques used in earlier tests. The peak or maximum deceleration test results were compared to the simulated analytical results. It was concluded that an analytical model based on DYNA3D code and has been adequately benchmarked for this type of application. A generic or represented cask was modified with the DYNA3D code and evaluated for ISFSI side drop and tipover events. The analytical method can be applied to similar casks to estimate impact loads on storage casks resulting from low-velocity side or tip impacts onto concrete storage pads.

Description:
This document presents the evaluation of the Spent Nuclear Fuel Cask and Transportation System. The system design was developed by Transnuclear, Inc. and its team members NAC International, Nelson Manufacturing, Precision Components Corporation, and Numatec, Inc. The cask is designated the TN-WHC cask. This report describes the design features and presents preliminary analyses performed to size critical dimensions of the system while meeting the requirements of the performance specification.

Description:
The Irradiated Fuel Storage Facility (IFSF) is a dry storage facility for spent nuclear fuels located at the Idaho Chemical Processing Plant; it was constructed in the 1970`s specifically for the Fort Saint Vrain spent reactor fuels. Currently, it is being used for various spent fuels. It was not known if IFSF would met current DOE seismic criteria, so re-analysis was started, with the rack being analyzed first. The rack was inspected to determine the as-built condition. LazrLyne and VideoRuler were used in lieu of using a tape measure with the camera. It was concluded that when a visual inspection shows widely varying weld sizes, the engineer has to use all resources available to determine the most probable specified weld sizes.

Description:
This report summarizes economic differences between a DUCRETE spent nuclear fuel storage cask and a conventional concrete storage cask in the areas of handling, transportation, and capital requirements. The DUCRETE cask is under evaluation as a new technology that could substantially reduce the overall costs of spent fuel and depleted U disposal. DUCRETE incorporates depleted U in a Portland cement mixture and functions as the cask`s primary radiation barrier. The cask system design includes insertion of the US DOE Multi-Purpose Canister inside the DUCRETE cask. The economic comparison is from the time a cask is loaded in a spent fuel pool until it is placed in the repository and includes the utility and overall US system perspectives.

Description:
The purpose of the WHC Systems Engineering Management Plan (SEMP) is to describe the systems engineering approach and methods that will be integrated with established WHC engineering practices to enhance the WHC engineering management of the SNF Project. The scope of the SEMP encompasses the efforts needed to manage the WHC implementation of systems engineering on the SNF Project. This implementation applies to, and is tailored to the needs of the SNF project and all its subprojects, including all current and future subprojects

Description:
Spent N-reactor fuel will be moved from wet to dry storage at Hanford Washington. The majority ofthis fuel exists as intact fuel assemblies, however, small amounts ofscrap will be included. Varying amounts of uranium metal are exposed in these fuel assemblies, depending upon the amount of mechanical damage sustained by the zircaloy cladding. The total exposed uranium surface area in each storage pool is estimated through the release of radioisotopes to the storage pools. The exposed uranium surface area of individual fuel assemblies in the K-East basin were estimated through the results of a camera survey. The exposed uranium surface area of scrap is estimated from the known particle side range and an estimated log normal particle size distribution. This document uses the radioisotope release calculations, the estimated scrap surface area, and the carnera survey results to estimate the ``worst case`` amount of surface area that could exist in a given ``MCO`` container containing 4 levels of fuel assemblies and one scrap basket. The total exposed uranium metal surface area for this ``worst case`` was 120,000 cm{sup 2}.

Description:
This paper describes experiments to determine the sodium retention and drainage from the two key areas of an ID-69. This information is then used as the initiation point for guidelines of how to proceed with washing an ID-69 in the IEM Cell Sodium Removal System.

Description:
This document establishes the functions and requirements baseline for the implementation of the Canister Storage Building Subproject. The mission allocated to the Canister Storage Building Subproject is to provide safe, environmentally sound staging and storage of K Basin SNF until a decision on the final disposition is reached and implemented

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